Loss of Coolant Accidents (LOCAs) is a crucial research topic for nuclear reactor safety analysis, and understanding the thermal–hydraulic behavior of the rod bundle channels during the reflooding stage of a LOCA is essential.

This study aims to develop theoretical models of the reflooding stage in addition to providing benchmark data for evaluating the safety analysis code for LOCAs in a reactor and for the design of the residual heat removal system.

A series of bottom reflooding tests were conducted on a 5×5 rod bundle in the film boiling test facility at the nuclear safety and operation laboratory (NUSOL) of Xi'an Jiaotong University using uniformly heated rods. The experimental results were analyzed in detail, and the surface parameters of the heated rod bundle were obtained by solving a one-dimensional transient inverse heat conduction problem. The effects of different experimental conditions on the velocity of the quench front propagation were investigated. Furthermore, the experimental results were compared and calculated using the thermal safety analysis code, and the problems with the thermal safety analysis code RELAP5 reflooding simulation are summarized.

Our results indicate that a high inlet flow rate, high inlet subcooling degree, and low power density are favorable for the propagation of the cold front during the reflooding process. Additionally, the root mean square (RMS) error of the simulated quench time and peak cladding temperature (PCT) are 40.994 s and 61.465 K, respectively. However, the simulation results have a relatively large error compared with the experimental results in the post-critical heat flux (CHF) heat transfer stage, primarily owing to the issues with the boiling mode judgment and membrane boiling heat transfer model.

The experimental data of this study can serve as new verification data for flow and heat transfer prediction models during the reflooding process; it can also be used to evaluate and optimize the thermal-hydraulic safety analysis code. Loss of Coolant Accidents (LOCAs) is a crucial research topic for nuclear reactor safety analysis, and understanding the thermal-hydraulic behavior of the rod bundle channels during the reflooding stage of a LOCA is essential.